This invention relates to reducing the corrosion potential of components exposed to high-temperature water. As used herein, the term "high-temperature water" means water of about 150.degree. C. or greater, steam, or the condensate thereof.
High-temperature water can be found in a variety of known apparatus, such as water deaerators, nuclear reactors, and in steam driven central station power generation.
Nuclear reactors are used in central-station electric power generation, research, propulsion, and for dual purposes thereof. A reactor pressure vessel contains the reactor coolant, i.e. water, which removes heat from the nuclear core. Piping circuits carry the heated water or steam to the steam generators or turbines, and circulated water or feedwater back to the vessel. Operating pressures and temperatures for the reactor pressure vessel are about 7 MPa and 288.degree. C. for a boiling water reactor, and about 15 MPa and 320.degree. C. for a pressurized water reactor. The materials used in both boiling water and pressurized water reactors must withstand various loading, environmental and radiation conditions.
Some of the materials exposed to the high-temperature water, steam, or condensate thereof include carbon steel, alloy steel, stainless steel, nickel-base alloys, and cobalt-base alloys. Additional information about such materials can be found in the Metals Handbook Desk Edition, American Society for Metals, 1985, with carbon and alloy steels at pp. 4.1-4.94, stainless steels and nickel-based alloys at pp. 15.1-15.30, and cobalt-base alloys at pp. 17.8-17.10, all incorporated herein by reference. Despite the careful selection and treatment of these materials for use in the systems, corrosion occurs on the materials exposed to the high-temperature water. Such corrosion contributes to a variety of problems, e.g., stress corrosion cracking, sticking of pressure relief valves, buildup of the gamma radiation emitting isotope .sup.60 cobalt, and erosion corrosion.
Stress corrosion cracking is a known phenomenon occurring in the components, including structural members, piping, fasteners, and weld deposits, of apparatus exposed to high-temperature water. For example, the components in nuclear reactors exposed to high-temperature water are known to undergo stress corrosion cracking. As used herein, the term "stress corrosion cracking" means cracking propagated by static or dynamic stressing in combination with corrosion at the crack tip. The reactor components are subject to a variety of stresses associated with, e.g., differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources including residual stress from welding, cold work and other asymmetric metal treatments. In addition, water chemistry, welding, heat treatment, and radiation can increase the susceptibility of a component to stress corrosion cracking of the metal. The factors that lead to stress corrosion cracking have been widely studied and a number of papers have been written concerning it.
Some of the publications addressing stress corrosion cracking and incorporated by reference herein are:
1) F. P. Ford, "Stress Corrosion Cracking", in Corrosion Processes, edited by R. N. Parkins, Applied Science Publishers, New York, 1982, p. 271.
2) J. N. Kass and R. L. Cowan, "Hydrogen Water Chemistry Technology for BWRs", in Proc. 2nd Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Monterey, Calif., 1985, p. 211.
3) M. E. Indig, B. M. Gordon, R. B. Davis and J. E. Weber, "Evaluation of In-Reactor Intergranular Stress" in Proc. 2nd Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Monterey, Calif., 1985, p. 411.
4) F. P. Ford, P. L. Andresen, H. D. Solomon, G. M. Gordon, S. Ranganath, D. Weinstein, and R. Pathania, "Application of Water Chemistry Control, On-Line Monitoring and Crack Growth Rate Models for Improved BWR Materials Performance", Proc. Fourth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Jekyll Island, Ga., August 1989, Nace, Houston, pp 4-26 to 4-51, 1990.
5) L. W. Niedrach and W. H. Stoddard, "Corrosion Potentials and Corrosion Behavior of AISI304 Stainless Steel In High Temperature Water Containing Both Dissolved Hydrogen and Oxygen", Corrosion, Vol 42, No. 12 (1986) page 696.
It is well documented that stress corrosion cracking occurs at higher rates when oxygen is present in the reactor water in concentrations of about 5 parts per billion, ppb, or greater. Stress corrosion cracking is further increased in a high radiation flux where oxidizing species, such as oxygen, hydrogen peroxide, and short-lived radicals are produced from radiolytic decomposition of the reactor water. Such oxidizing species increase the electrochemical corrosion potential of metals. Electrochemical corrosion is caused by a flow of electrons from anodic and cathodic areas on metallic surfaces. The corrosion potential is a measure of the thermodynamic tendency for corrosion phenomena to occur, and is a fundamental parameter in determining rates of, e.g., stress corrosion cracking, corrosion fatigue, corrosion film thickening, and general corrosion.
As explained in these and other articles, stress corrosion cracking in boiling water nuclear reactors and the associated water circulation piping has been reduced by injecting hydrogen in the water circulated therein. The injected hydrogen reduces oxidizing species in the water, such as dissolved oxygen, and as a result lowers the corrosion potential of metals in the water. However, factors such as variations in water flow rates and the time or intensity of exposure to neutron or gamma radiation result in the production of oxidizing species at different levels in different reactors. Thus, varying amounts of hydrogen have been required to reduce the level of oxidizing species sufficiently to maintain the corrosion potential below a critical potential required for protection from the stress corrosion cracking in the high-temperature water.
As used herein, the term, "critical potential" means a corrosion potential at or below a range of values of about -230 to -300 mV based on the standard hydrogen electrode (she) scale. Below the critical potential, stress corrosion cracking is markedly reduced or even eliminated as disclosed in references 2-5. Stress corrosion cracking proceeds at an accelerated rate in systems in which the electrochemical potential is above the critical potential, and at a substantially lower rate in systems in which the electrochemical potential is below the critical potential. Water containing oxidizing species such as oxygen increases the corrosion potential of metals exposed to the water above the critical potential, while water with little or no oxidizing species present results in corrosion potentials below the critical potential.
Corrosion potentials of stainless steels in contact with reactor water containing oxidizing species can be reduced below the critical potential by injection of hydrogen into the water in a concentration of about 50 to 100 ppb or greater. Much higher hydrogen injection levels are necessary to reduce the corrosion potential within the high radiation flux of the reactor core, or when oxidizing cationic impurities, e.g., cupric ion are present. Such hydrogen injection lowers the concentration of dissolved oxidizing species in the water and also the corrosion potential of the metal. However, high hydrogen additions, for example of about 150 ppb or greater, that reduce the corrosion potential below the critical potential can result in a higher radiation level in the steam driven turbine section from incorporation of the short-lived N.sup.16 species. The higher radiation requires additional shielding, and radiation exposure control.
Accordingly, although the addition of hydrogen lowers the corrosion potential of reactor water, it is also desirable to limit the amount of hydrogen in reactor water, while maintaining the corrosion potential below the critical potential.
Two disclosures dealing with the formation of noble metal deposits on stainless steel are:
(6) H. Ocken, C. C. Lin, and D. H. Lister, "Thin Films to Impede the Incorporation of Radio Nucleides in Austenetic Stainless Steels", Thin Solid Films, Vol. 171 (1989) pages 323-334.
(7) G. P. Chernova, T. A. Fedosceva, L. P. Kornienko, and N. D. Tomashov, "Increasing Passivation Ability and Corrosion Resistance of Stainless Steel by Surface Alloying with Palladium", Prot. Met. (Eng. Transl.) 17 (1981) page 406.
The first of these disclosures deals with the use of pre-oxidation treatments and metallic films to reduce the incorporation of .sup.60 Co in components of the circulatory system of a nuclear reactor that are in contact with the coolant. An electroless deposit of palladium was found to be most effective in reducing incorporation of .sup.60 Co in stainless steel exposed to simulated primary water coolant conditions, i.e., water that is not injected with hydrogen.
The second disclosure deals with the electrochemical behavior and increase in corrosion potential and passivation of stainless steel that is surface alloyed with palladium, and exposed to deaerated acidic solutions at room temperature. The increased corrosion potential causes a passive oxide layer to form on the stainless steel that reduces further corrosion.
One object of this invention is to provide a method for reducing corrosion of a component exposed to high-temperature water.
Another object is to provide a method for reducing stress corrosion cracking of a component exposed to high-temperature water.
Another object is to provide a method for reducing corrosion by lowering the corrosion potential of a component exposed to high-temperature water.
Another object is to provide a method for reducing the amount of hydrogen injected into high-temperature water to reduce the corrosion potential of a component exposed to high-temperature water.